ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for the K-25 Site – Occupational 

Environmental Dose

Document Number:

ORAUT-TKBS-0009-4 

Effective Date: 09/26/2006

Revision No.: 00 PC-1 

Controlled Copy No.: ________

Page 1 of 21

Subject Expert: James East

Document Owner

Approval: Signature on File Date: 12/20/2004 Jay J. Maisler, TBD Team Leader

Approval: Signature on File Date: 12/20/2004 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 12/20/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 12/29/2004

James W. Neton, Associate Director for Science

Supersedes:

None

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

3

Acronyms and Abbreviations ...............................................................................................................5

4.1 Introduction................................................................................................................................6

4.2 Internal Dose from Onsite Atmospheric Radionuclide Concentrations ....................................... 7

4.2.1 Onsite Releases to Air ......................................................................................................7

4.3 External Dose..........................................................................................................................11

4.3.1 Ambient Radiation .......................................................................................................... 11

4.3.2 Other Potential Sources.................................................................................................. 12

4.3.3 Review of External Dose................................................................................................. 12

4.4 Uncertainty ..............................................................................................................................12

References ........................................................................................................................................14

Glossary ............................................................................................................................................15

Attachment 4D Occupational Environmental Dose ...........................................................................17

4D.1 Internal Dose from Onsite Concentrations of Radionuclides .................................................... 18

4D.1.1 Onsite Releases to Air .................................................................................................... 18

4D.2 External Dose..........................................................................................................................18

4D.2.1 Ambient Radiation .......................................................................................................... 18
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LIST OF TABLES

Table

4D-1 Uranium annual intake per year (Bq) and geometric standard deviation................................. 

Page

19

4D-2 Plutonium annual intake per year (Bq) and geometric standard deviation .............................. 20

4D-3 Neptunium and technetium annual intake per year (Bq) and geometric 

standard deviation..................................................................................................................21

4D-4 Background gamma radiation levels at two perimeter ORGDP stations ................................. 21

LIST OF FIGURES

Figure

4-1 K-25 Site map with environmental monitoring stations noted ...................................................

Page

8
Effective Date: 09/26/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0009-4 Page 3 of 21

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE REV. NO. DESCRIPTION

Draft 09/26/2003 00-A New Technical Basis Document for the K-25 

Site – Occupational Environmental Dose. 

Initiated by Joseph L. Alvarez.

Draft 10/22/2003 00-B Incorporates comments from NIOSH review. 

Initiated by Joseph L. Alvarez.

Draft 01/26/2004 00-C Incorporates additional comments from 

NIOSH review. Initiated by Jay J. Maisler.

Draft 05/21/2004 00-D Incorporates additional comments from 

NIOSH review. Initiated by Jay J. Maisler.

Draft 08/17/2004 00-E Incorporates additional comments from 

NIOSH review. Initiated by Jay J. Maisler.

Draft 12/13/2004 00-F Incorporates additional comments from 

NIOSH review. Initiated by Jay J. Maisler.

12/29/2004 12/29/2004 00 First approved issue. Initiated by Jay J. 

Maisler.

12/29/2004 09/26/2006 00 PC-1 Approved page change revision to include 

new standard language on page 6 in Section 

4.1. Adds reference to page 13 in Reference 

Section. No sections were deleted. No 

changes occurred as a result of internal formal 

review. 

OCAS comment was addressed. No changes 

were required due to the K25 Worker 

Outreach SPFPA and IGUA Worker Outreach 

meeting held on April 5, 2006 or K25 Worker 

Outreach PACE Local 5-288 and Pace 

International Union Worker Outreach meeting 

held on April 6, 2006. However, the document 

previously addressed the following comments:

From K25 Worker Outreach SPFPA and IGUA 

Worker Outreach meetings April 5, 2006- 

Response to Comment #30: Atmospheric 

releases are addressed in Section 4.2.1.

From K25 Worker Outreach PACE Local 5-

288 and Pace International Union Worker 

Outreach comments April 6, 2006- 

Response to Comment #4: Fires are 

discussed in the reference "DOE (U.S. 

Department of Energy), 2000b, Independent 

Investigation of the East Tennessee

Technology Park." 

Response to Comment #33: Atmospheric 

releases of Uranium and other radioactive 

materials are addressed in Section 4.2.1
Effective Date: 09/26/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0009-4 Page 4 of 21

Response to Comment #45: Because of the 

known limitations of air monitoring 

environmental intakes are calculated based on 

releases of materials; this is discussed in 

Section 4.2.

Response to Comment #46: As documented 

in Section 4.3.1, neutron exposures were not 

reported for environmental results until 2003.

This revision results in no change to the 

assigned dose and no PER is required. 

Training required: As determined by the Task 

Manager. Initiated by Paul A. Szalinski. 

Approval:

Paul A. Szalinski, Document Owner

Signature on File 08/31/2006

John M. Byrne, Task 3 Manager

Signature on File 09/05/2006

Signature on File ______

Edward F. Maher, Task 5 Manager

08/31/2006

Kate Kimpan, Project Director

Signature on File 09/14/2006

James W. Neton, Associate Director for Science

Brant A. Ulsh Signature on File for 09/26/2006
Effective Date: 12/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-4 Page 5 of 21

ACRONYMS AND ABBREVIATIONS

DOE U.S. Department of Energy

GSD Geometric Standard Deviation

ORDR Oak Ridge Dose Reconstruction

ORHASP Oak Ridge Health Agreement Steering Panel

ORGDP Oak Ridge Gaseous Diffusion Plant

ORR Oak Ridge Reservation

RU recycled uranium

TLD thermoluminescent dosimeter

U.S.C United States Code
Effective Date: 09/26/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0009-4 Page 6 of 21

4.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historical background information and guidance to assist in the preparation of 

dose reconstructions for particular sites or categories of sites. The documents will be revised in the 

event additional relevant information is obtained about the affected site(s). These documents may be 

used to assist NIOSH in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 

U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures 

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Effective Date: 12/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-4 Page 7 of 21

The occupational environmental dose refers to the dose received by unmonitored workers outside the 

facilities. These doses can be internal and external depending on the characteristics of the individual 

radionuclides. When inhaled, most radionuclides would produce a dose to various organs in the 

body. These radionuclides are addressed in the following sections. 

4.2 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

The internal dose for workers outside the facilities is determined from the air concentrations resulting 

from individual facility releases, and from ground-level releases. The method used to predict air 

concentrations includes the contribution from soil resuspension indirectly. 

Unmonitored workers can be exposed to occupational doses internally from onsite releases to the air 

and from the resuspension of radioactive materials in soil, and externally from ambient radiation and 

releases of radioactive noble gases to the air.

4.2.1

The internal dose for workers outside the facilities is determined from the air concentrations resulting 

from releases from stacks, individual building releases, and from the purge cascade and other 

operations at the Oak Ridge Gaseous Diffusion Plant Site (ORGDP), which this document refers to as 

the K-25 Site. Unmonitored workers can be exposed to occupational doses internally from onsite 

releases to the air or from ambient radiation.

Onsite Releases to Air

A calculation was performed using well-documented source terms developed for the Oak Ridge Dose 

Reconstruction (ORDR) (Burmeister 1996, 1997, Burns 1997), coupled with some documented 

environmental monitoring data, to estimate radionuclide-specific airborne concentrations for 234, 235, 

238U, 238, 239, 240, 241, 242Pu, 237Np, and 99Tc (Shonka 2003). These radionuclides were determined to 

account for the majority of the potential missed dose from inhalation and submersion pathways. The 

50th- and 95th-percentile air concentrations for the uranium nuclides were used to compute the 

geometric standard deviation (GSD) of the uranium source term. This GSD was applied on a year-byyear basis to all other nuclides in the source term. The 50th-percentile air concentration for uranium 

calculated air concentrations was then used to derive annual intakes based on an assumed individual 

ventilation rate of 2,400 cubic meters per year. In addition, the GSD of these intakes was estimated 

based on the 95th-percentile source terms using the following relationship:

th

th

1

Nuclide activity 95 percentile 1.65585

 = Nuclide activity 50 percentileGSD

      

   

These annual intakes and GSDs, by year of operation, are listed in Attachment 4D in Tables 4D-1 

through 4D-3. 

In general, the K-25 Site was a major industrial site with upwards of 2,500 megawatts of heat 

dissipated in the water and air during the years of operation. Releases from the gaseous diffusion 

plants had both mechanical momentum and significant thermal energy that lifts the effective height of 

the release point by 10 to 20 meters. Depending on the location of a specific release point relative to 

other turbulent thermal sources such as cooling towers, the diffusion can be considerable over that 

predicted by simple Gaussian plume models. There were multiple release points spread over the site. 
Effective Date: 12/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-4 Page 8 of 21

Figure 4-1 is a drawing of the K-25 Site showing major buildings and the environmental stations 

selected to estimate dilution.

Figure 4-1. K-25 Site map with environmental monitoring stations noted. 

To provide a conservative estimate, environmental monitors were chosen that were adjacent to the 

K-25 Site and generally located in upwind and downwind locations relative to release points. These 

were where maximum ground-level concentrations from the site would occur. The annual air 

concentrations for several years in gross alpha and gross beta were divided by the source term for the 

same years expressed in the same units to provide an equivalent annual dilution volume. This 

dilution volume was then applied to all years. Use of gross values provides added conservatism in 

that nearly all of the reported air concentration appears to be background. In addition, the 

administration area is generally crosswind from the bulk of the industrial site where most of the 

release points were located.

The values in Tables 4D-1 through 4D-3 in Attachment 4D are estimates of annual intakes derived 

from the calculated outdoor air concentrations at the K-25 Site and, as such, might not be appropriate 

for calculating organ doses for workers located inside buildings. However, these values probably

represent the upper bound of potential intakes for an individual working in outside areas where 
Effective Date: 12/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-4 Page 9 of 21

individual monitoring was not required. It is assumed that these intakes occurred uniformly in the year 

listed, so for partial years of employment the intake can be proportionally adjusted by the fraction of 

the year worked. In addition, these intakes were derived based on an assumed breathing rate of 

2,400 cubic meters per year. Therefore, adjustments might be required based on the job 

classification depending on the individual performing light work or heavy work, for which the 

respiration rate would be 1.2 m3

/h or 1.7 m3/h, respectively. For purposes of dose estimation, 

radioactive decay was not considered.

Plutonium and uranium releases to the atmosphere represent the greatest potential radiological risk of 

all the alpha-emitting radionuclides in the K-25 Site environment. For purposes of dose 

reconstruction, the isotopes of plutonium and uranium contained in the mix that results in the 

maximum organ dose should be selected and assumed to be present at 100%. This assumption will 

result in a small, but claimant-favorable, overestimation of the actual dose.

Releases of radionuclides for the K-25 Site were studied as a part of the ORDR. This study was 

conducted from 1994 to 1999 for the Tennessee Department of Health by scientists and engineers 

from ChemRisk, a service of McLaren/Hart Environmental Services, assisted by SENES Oak Ridge 

and Shonka Research Associates. As an element of Task 6 of that study, a source term 

reconstruction was performed for the primary nuclides of concern, that of the uranium nuclides at 

various enrichments. In addition, a screening source term was developed for 237Np and 99Tc as an 

element of Task 7 of the ORDR. The screening source term used plant data and did not 

independently reconstruct the release. These source terms were used as the starting point for this 

effort. 

There were two uranium source terms developed for K-25 by ORDR. The final source term reported 

in the Task 6 Final Report did not have 50th- and 95th-percentile values, and several limitations of the 

work were noted (ChemRisk 1999). These included limited release information for the S-50 plant for 

1944, the sole year of operation for that facility. In addition, the K-25 Site releases did not resolve 

releases from K-1131 against the large material unaccounted for (mass balance) for that building. 

Both of those source terms are primarily natural enrichment material of low specific activity. In 

addition, undocumented releases from the cascade were noted from frequent incident reports that had 

no quantity release data associated with the incident. Finally, corrections for losses from stack 

sampling were not rigorously treated. One of the primary stack monitors whose results would be 

impacted is the K-1131 stack, which is primarily a source of near-natural enrichment uranium. 

The alternative uranium release estimate was an attempt to address some of the issues mentioned 

above, and provided 50th- and 95th-percentile estimates. It estimated uncertainty by fitting release 

information for periods when it was available, and sampling from the fitted data to fill in periods where 

monitoring data was unavailable. The Oak Ridge Health Agreement Steering Panel (ORHASP) felt 

that the conclusions reached were unchanged for either of these treatments, and preferred to use the 

data without fits, as it reflected the reported plant releases. The uranium release from the two 

estimates was 16.7 curies (Burns 1997) and 23.2 curies (Burmeister 1996) (50th percentile), 

respectively. 

The screening estimate made for ORDR Task 7 took the inventory of 99Tc and 237Np estimated by the 

K-25 Site staff and the Site’s release fraction and distributed that material throughout the years of 

recycled uranium (RU) use by establishing the ratio of uranium releases for that period. After 1977, a 

nominal continuing annual release was used. No estimate was made for plutonium in that Task 7 

effort. 
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To provide a conservative value that was claimant-friendly, the 50th-percentile alternative release 

estimate from the ORDR source term was used for uranium, and the 95th-percentile was used to 

calculate the geometric standard deviation along with the following modifications. Plutonium was 

added in a manner consistent with the previous estimates for 99Tc and 237Np, and with an assumption 

that the material consisted of five plutonium nuclides in a ratio consistent with weapons-grade 

plutonium. 

Estimates of site releases for 99Tc were used for 1981 and 1982 in lieu of the ORDR estimates 

because the reported value was larger than the estimate (Jordan 1983). On average, this serves to 

overstate the 99Tc releases, since the added curies (beyond those asserted by ORDR Task7) had 

been distributed by the ORDR calculation among all years of RU use.

The K-25 Site is relatively small compared to larger U.S. Department of Energy sites, and radioactive 

material is released from a large number of release points of varying flow rates and elevations 

(Shonka 1996). Large quantities of electric power were used (depending on production), with heat 

rejected, in part, to the atmosphere using a combination of ventilation and cooling towers. Ventilation 

rates for removal of process heat varied to a considerable degree with the seasons. This, along with 

building wake effects, result in a highly turbulent environment that tends to mix and dilute the released 

materials far beyond the rates predicted by atmospheric transport codes. In light of the multiple 

release points and complexities of transport, a simple conservative method for estimating site 

concentrations was chosen.

The gross alpha and gross beta boundary air concentrations, reported by the site in environmental 

monitoring reports from the 1970s and 1980s, were ratioed to the source term (expressed in 

equivalent units) to establish a dilution volume that was used for all years. The reported boundary air 

concentrations were nearly identical to those recorded by offsite monitors, indicating that the 

concentrations came primarily from sources that were not associated with releases from K-25 

operations. Thus, this method provides an estimate of annual concentrations that is overstated and 

conservative. This excess conservatism is used to account for the limitations in the source term noted 

above as well as other airborne sources including soil resuspension. No separate analysis for soil 

resuspension was performed.

Once the site boundary intakes are established, the intakes were scaled back to within the fence line 

using Fig 11.3.4 in the Radiological Health Handbook

In the concentration factor, the average wind speed (ū) between the boundary location and the fence 

line would be the same, and therefore isn’t a contributing factor in relating the two distances. Also the 

release (Q) is the same in both cases, leaving the only change in concentration (χ).

(Shleien 1992). From this figure based on 

ground level releases the concentration factor (ū χ/Q) to distance (x) ratio is -102 to 101 is evident. 

Ground level releases provides for the highest intake in the close distances.

The boundary concentrations, as reported in referenced calculations and reports, were taken at 

approximately 2500 meters from the center of the plant, the fence line is generally less than 100 

meters from release points. Therefore, reducing the distance by 1.5 magnitudes to 80 m increases 

the intake within the fence line by a magnitude of three over the site boundary intake. Intakes from 

the reference calculations have been increased by a factor of 103 and tabulated in the attachments 

and should be used for unmonitored environmental intake.
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4.3 EXTERNAL DOSE

Workers are subjected to external doses from the ambient radiation levels at the K-25 Site. 

Environmental data was available only from 1973 through 1985. These data was collected from CaF 

thermoluminescent dosimeters (TLDs) mounted 1 meter above the ground. The TLDs were read 

monthly for two K-25 Site perimeter stations: HP-33 (Gallagher Gate) and HP-35 (Blair Gate). 

Perimeter monitoring stations are not the best source of data for onsite exposure for unmonitored 

workers. Unlike PORTS and PGDP, records have not revealed a network of environmental 

dosimeters along the fence line or within the K-25 site. This may be attributed to the practice of 

including K-25 in environmental reports for the entire Oak Ridge Reservation (ORR) which also 

includes the Y-12 and ORNL facilities. K-25 was perceived to have little external exposure impact in 

comparison with other ORR facilities and the monitoring was focused on the area most affected 

around ORNL. The lesser impact is indeed the case at K-25, but there is little real world data to base 

such a conclusion. What remains is the manual surveys and environmental placements at sister 

facilities to draw from.

4.3.1

Site annual environmental reports were reviewed for data that would be useful in reconstructing 

ambient radiation levels and airborne radionuclide concentrations. Data in these documents included 

ambient TLD radiation measurements at perimeter stations. The historical TLD results with 

uncertainty estimates, where available, were tabulated and are presented in Table 4D-4. An asterisk 

indicates years with data, other years are estimated from these data. 

Ambient Radiation

The ambient radiation levels listed in Table 4D-4 were taken directly from the K-25 Site annual and 

semiannual environmental reports and include contributions from K-25 Site releases and background 

radiation. The values in Table 4D-4 are average readings in μR/hr from a month-long exposure 

period. All of these values include the contributions from natural background radiation and nuclear 

weapons fallout. Therefore, use of these data to account for potential missed or unmonitored dose 

should in all cases represent an overestimation of the worker’s actual dose. 

For years that have no reported data, the average of the two stations used is reported for the 50th

percentile. The GSD was computed via the equation provided in Section 4.1.1 of this chapter. One 

data point (1982, HP35: Blair Gate) was not used for computing the GSD for years with no data, 

because those data are a significant outlier compared to the other data. 

Uranium cylinder storage yards remain the only significant source of external exposure at K-25. 

Surveys in cylinder yards at the sister plants show dose rates up to 200 mrem/2000 h. For 

simplification, it can be assumed that the cylinder yard inventory grew linearly over time during the 

production years. 

The ORR Annual Environmental Report for 2003 (DOE 2004) reports a dose from the K-25 cylinder 

yards to a nearby parking area at 4.75 mrem/125h (75 mrem/2000h). The 2003 report was the first to 

include neutron exposure, which represents 63% of the exposure. The dose at the parking lot was 

only reported in two other years, but these did not include the neutron component.

Though this dose rate could not have existed in the early years when DU inventory was just 

beginning, it should be applied to unmonitored workers for each full year of employment at K-25. This 

would conservatively assign more dose than can be established from the record.
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4.3.2

Other sources of photon exposure have been investigated in recent years (DOE 2000a,b). Both of 

these documents investigated doses from fission products and found such doses to be insignificant. 

These reports also indicated that K-25 did investigate processing of 137Cs waste but neither report

found indications of significant dose to individuals from the investigation.

Other Potential Sources

Current levels of fission products as surface and other contamination were also investigated (DOE 

1984). Only 137Cs was found in trace amounts in 3 of 96 samples taken in 19 buildings. Such 

evidence shows that 137Cs and other photon emitters need not be considered as sources of exposure. 

4.3.3

The mean dose rate for the region of K-25 is 7.8 μR/hr (15.6 mR/2000h) (DOE 2001). The mean 

dose rate used in the evaluation is 8.6 μR/h (17.2 mR/2000h). The assumption that the entire 

recorded environmental dose is attributed to K-25 is very conservative because the difference is 0.8 

μR/hr. The airborne releases in Tables 4D-1 through 4D-3 show that releases were comparable in 

1944-1973 to those of 1973-1985. There is no indication based on releases that dose estimates 

should be different. ORAU (2004) found upper bound dose rates for unreported doses for personnel 

in buildings was 25 μR/hr (50 mR/2000h) for the years in question. It is unlikely that environmental

doses approached this level outside the buildings, except near the cylinder yards. The use of the 

mean dose rate for K-25 during 1973-1985 as an estimate for 1944-1973 is conservative because the 

majority of the estimated dose rate is from natural background. The further extension of the estimate 

to the 95th percentile increases that conservatism.

Review of External Dose

Because 12.58 μR/hr (25.2 mR/2000h) is the 99th percentile of the mean dose rate, which is primarily 

natural background, the true dose from K-25 operations is more than confidently contained in the 

estimate. 

The recommended environmental external dose is 90 mRem/y. This is based on the sum of the mean 

dose rate for the region (15.6 mRem) and the dose from DU Cylinder Yards (75 mRem). Other 

sources of environmental external dose need not be considered.

4.4 UNCERTAINTY

As discussed in the previous sections, estimates of annual intakes employed conservative (i.e., 

claimant-favorable) assumptions. The intake values in Attachment D should represent a reasonable 

upper bound of the actual intakes that could have occurred. In instances where more detailed 

information is known about a particular individual or job classification, other modifying factors could be 

employed. For example, if, for a particular job classification, there is reason to believe that the actual 

breathing rate for the claimant might vary markedly from the average of 2,400 cubic meters per 2,000 

hours of exposure, the dose reconstructor should use professional judgment to adjust the estimated 

intakes as necessary whether the individual is engaged in light or heavy work. The respiration rate in 

these cases for light work is 1.2 m3

/hr and 1.7 m3/hr for heavy work. 

To estimate the annual intake in these cases, the product of the fractional annual period for each jobdependent level of work and the corresponding ventilation rate are summed to determine the total 

ventilation volume for the year in cubic meters. The annual intake is then determined by multiplying 

the annual ventilation volume by the annual average concentration in Tables 4D-1 through 4D-3 for 

the year and location of interest.
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In general, the annual intake values presented in Tables 4D-1 through 4D-3 reflect adjusted source 

terms and uncertainty analyses provided in the ORDR estimate (ChemRisk 1999). For purposes of 

this document, no other attempt has been made to quantify other uncertainties. 
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REFERENCES

Burmeister, R. E., 1996, The Atmospheric Master Release List and Atmospheric Source Term for 

K-25, SRA 96-012. 

Burmeister, R. E., 1997, 237Np and 99Tc Released to Air and Water from Historic Operations at the 

K-25 Site and Y-12 Sites, SRA-97-003. 

Burns, R. E., 1997, Per Annum Atmospheric Releases of Uranium from the K-25 Site from 1944 to 

1995, SRA-97-011. 

ChemRisk, 1999, Reports of the Oak Ridge Dose Reconstruction, Vol. 5, The Report of Project Task 

6, July 1999, Uranium Releases from the Oak Ridge Reservation – a Review of the Quality of 

Historical Effluent Monitoring Data and a Screening Evaluation of Potential Off-Site Exposures. 

DOE (U.S. Department of Energy), 1984, Assessment of Accessible Contamination at the K-25 Site 

Phase 3 Report: Cumulative Analytical Results, K/HS-570. 

DOE (U.S. Department of Energy), 1992, An Aerial Radiological Survey of the Oak Ridge Reservation 

and Surrounding Area, EGG-10617-1123, UC-702. 

DOE (U.S. Department of Energy), 2000a, Recycled Uranium Mass Balance Project Oak Ridge 

Gaseous Diffusion Plant, BJC/OR-584. 

DOE (U.S. Department of Energy), 2000b, Independent Investigation of the East Tennessee 

Technology Park, Office of Oversight Environment, Safety, and Health, Washington, D.C. 

DOE (U.S. Department of Energy), 2001, Oak Ridge Reservation Environmental Report for 2000, 

DOE/ORO/2119, Oak Ridge, Tennessee. 

DOE 2003, Oak Ridge Reservation Environmental Report for 2003, DOE/ORO/2186 

http://www.ornl.gov/sci/env_rpt/aser2003/aser2003.html

Jordan R.G., 1983, “Radioactive Effluent Data,” letter.

ORAU (Oak Ridge Associated Universities), 2004, Technical Basis Document for the K-25 Site – 

External Dose, ORAUT-TKBS-0009-6, Oak Ridge, Tennessee.

Shleien, B, et. al., 1992, Handbook of Health Physics and Radiological Health.

Shonka, J., 1996, “Emission Point Characteristics," Project Memorandum JJS.050 to J. Buddenbaum. 

Shonka, J., 2003, Calculations for the ORGDP Technical Basis Document (TBD) – Modified 

Atmospheric Master Routine List for ORGDP and Background Gamma Radiation, SRA 03-

001.
Effective Date: 12/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-4 Page 15 of 21

GLOSSARY

background radiation 

The radiation received by man from natural and environmental sources including cosmic rays, 

radiation from the naturally radioactive elements in the environment, and natural 

concentrations of radionuclides in the body (carbon-14, potassium-40). The usually quoted 

average individual exposure from background radiation for a person living in the continental 

United States is 250 to 300 millirem per year. 

depleted uranium 

On the ORR, depleted uranium consisted mostly of U-238 and usually contains between 0.14 

and 0.20% uranium-235 by weight. Natural uranium contains 0.72% uranium-235, while 

enriched uranium contains greater than 0.72% uranium-235 by weight. For example, depleted 

uranium is generated as a result of the K-25 gaseous diffusion uranium enrichment and is 

found in the tailings portion of the process outputs. 

enriched uranium 

On the ORR, enriched uranium typically contained between 0.95% and 99% uranium-235. 

Natural uranium contains 0.72% uranium-235, while depleted uranium contains less than 

0.72% uranium-235. 

enrichment of uranium 

A process in which the relative abundance of one of the isotopes of uranium is increased with 

respect to the others. These processes in the past used the difference in the mass of the 

isotopes to increase the relative fraction of one isotope over the others. The resultant material 

is enriched in one particular isotope (usually uranium-235) and depleted in its other isotope 

(uranium-238). 

gaseous diffusion enrichment 

A process by which uranium hexafluoride is passed through a series of semipermeable 

molecular barriers for the purpose of separating the lighter uranium-235 isotope from the 

heavier, more naturally abundant uranium-238 isotope (see enrichment of uranium). 

natural uranium 

Natural or “normal” uranium contains 0.72% uranium-235. Contrast with enriched uranium, 

which contains more than the natural concentration of uranium-235, and depleted uranium, 

which contains less than 0.72% uranium-235. 

percentiles 

If a large set of data is arranged from its smallest value to its largest, and this list is divided into 

100 classes containing nearly equal numbers of data, then each percentile represents the 

highest value within that class. Thus 5% of the data are less than or equal to the 5th 

percentile, and approximately 95% of the data are greater than or equal to the 5th percentile. 

The median is defined as the 50th percentile, which divides the data (approximately) into 

halves. 

source term 

The quantity, chemical and physical form, and time history of contaminants released to the 

environment from a facility. 
Effective Date: 12/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-4 Page 16 of 21

uncertainty 

The level of confidence in a given estimate based on the quality and quantity of the available 

data. Inherent uncertainties are generated by a number of sources including uncertainties in 

measurements, absence of data due to the lack of environmental monitoring, lack of 

knowledge about some physical processes and operational procedures, and the approximate 

nature of mathematical models used to predict the transport of released materials. 

uranium 

A naturally occurring radioactive metal that, in natural ores, has an atomic weight of 

approximately 238. The two principal natural isotopes are uranium-235 (0.7%) and uranium238 (99.3% of natural uranium). Natural uranium also includes a very small amount of the 

daughter uranium-234 by weight. The activity associated with this U-234 is significant as U235 enrichment increases. Uranium has been used chiefly in nuclear reactors and nuclear 

explosives.
Effective Date: 12/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-4 Page 17 of 21

ATTACHMENT 4D

OCCUPATIONAL ENVIRONMENTAL DOSE

TABLE OF CONTENTS

4D.1 Internal Dose from Onsite Concentrations of Radionuclides .................................................... 18

 4D.1.1 Onsite Releases to Air............................................................................................... 18

4D.2 External Dose..........................................................................................................................18

 4D.2.1 Ambient Radiation ...................................................................................................... 18

LIST OF TABLES

4D-1 Uranium annual intake per year (Bq) and geometric standard deviation................................. 19

4D-2 Plutonium annual intake per year (Bq) and geometric standard deviation .............................. 20

4D-3 Neptunium and technetium annual intake per year (Bq) and geometric 

standard deviation..................................................................................................................21

4D-4 Background gamma radiation levels at two perimeter ORGDP stations ................................. 21
Effective Date: 12/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-4 Page 18 of 21

ATTACHMENT 4D

OCCUPATIONAL ENVIRONMENTAL DOSE

Unmonitored site workers could be exposed to occupational doses internally from the onsite radiation 

releases to the air and externally from ambient radiation.

4D.1 INTERNAL DOSE FROM ONSITE CONCENTRATIONS OF RADIONUCLIDES

4D.1.1

If a worker spent significant time in areas for which radionuclide concentrations were calculated and 

the worker was not included in the internal dosimetry program, the data in Tables 4D-1 through 4D-3 

can be used to estimate unmonitored intakes. These tables list average derived intake (assuming a 

2,400-cubic-meter-per-year ventilation rate) at the 50th percentile and the GSD based on the 95th

percentile. For plutonium, neptunium, and technetium the GSD was computed from the total uranium. 

The decision to use these air concentrations to estimate unmonitored intake should be made 

cautiously because the worker’s intake could be significantly overestimated.

Onsite Releases to Air

4D.2 EXTERNAL DOSE

4D.2.1

The ambient radiation levels in Table 4D-4 are listed in μR/hr for various perimeter locations at the 

ORGDP. The values denoted with a “∗” are derived from empirical Site data. The other values are 

computed from the empirical Site data and propagated to the other years of Site operations. The 

empirical values in Table 4D-4 include contributions from natural background radiation and nuclear 

weapons fallout. Therefore, use of these data to account for potential missed or unmonitored dose 

should, in all cases, represent an overestimation of the worker’s actual dose. 

Ambient Radiation
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Table 4D-1. Uranium annual intake per year (Bq) and geometric 

standard deviation. 

Date

U-234 U-235 U-238

50th percentile GSD 50th percentile GSD 50th percentile GSD

1944 5.57E-02 Bq 1.99 2.65E-03 Bq 1.99 5.76E-02 Bq 1.99

1945 1.69E-01 Bq 1.98 8.03E-03 Bq 1.98 1.74E-01 Bq 1.98

1946 6.34E-02 Bq 1.6 3.05E-03 Bq 1.6 1.58E-02 Bq 1.6

1947 6.99E-02 Bq 1.67 3.16E-03 Bq 1.6 1.57E-02 Bq 1.6

1948 6.75E-02 Bq 1.69 3.06E-03 Bq 1.6 1.64E-02 Bq 1.6

1949 7.44E-02 Bq 1.64 3.36E-03 Bq 1.6 2.66E-02 Bq 1.6

1950 8.50E-02 Bq 1.6 3.89E-03 Bq 1.6 3.46E-02 Bq 1.6

1951 8.49E-02 Bq 1.6 6.64E-03 Bq 1.6 9.84E-02 Bq 1.6

1952 1.74E-01 Bq 1.6 7.76E-03 Bq 1.6 1.69E-01 Bq 1.6

1953 3.09E-01 Bq 1.6 1.56E-02 Bq 1.6 1.81E-01 Bq 1.6

1954 3.36E-01 Bq 1.6 1.39E-02 Bq 1.6 6.24E-02 Bq 1.6

1955 9.83E-02 Bq 1.83 5.49E-03 Bq 1.6 6.70E-02 Bq 1.6

1956 1.30E-01 Bq 1.96 5.84E-03 Bq 1.6 5.49E-02 Bq 1.6

1957 1.43E-01 Bq 1.89 6.44E-03 Bq 1.6 6.13E-02 Bq 1.6

1958 4.82E-01 Bq 1.6 2.31E-02 Bq 1.6 3.78E-01 Bq 1.6

1959 2.64E-01 Bq 1.6 1.14E-02 Bq 1.6 9.31E-02 Bq 1.6

1960 2.01E-01 Bq 1.6 9.65E-03 Bq 1.6 1.66E-01 Bq 1.6

1961 1.82E-01 Bq 1.6 8.08E-03 Bq 1.6 1.25E-01 Bq 1.6

1962 1.16E-01 Bq 1.6 5.39E-03 Bq 1.6 2.22E-02 Bq 1.6

1963 1.93E+00 Bq 1.6 1.02E-01 Bq 1.6 1.25E-01 Bq 1.6

1964 3.58E-02 Bq 2.89 1.59E-03 Bq 1.91 3.07E-03 Bq 1.78

1965 3.00E-01 Bq 1.6 1.59E-02 Bq 1.6 6.17E-02 Bq 1.6

1966 3.40E-02 Bq 2.97 1.51E-03 Bq 1.95 2.49E-03 Bq 1.93

1967 3.38E-02 Bq 2.98 1.50E-03 Bq 1.96 2.41E-03 Bq 1.96

1968 3.39E-02 Bq 2.97 1.50E-03 Bq 1.95 2.66E-03 Bq 1.88

1969 3.53E-02 Bq 2.91 1.57E-03 Bq 1.92 3.81E-03 Bq 1.65

1970 3.56E-02 Bq 2.9 1.60E-03 Bq 1.91 3.24E-03 Bq 1.75

1971 6.40E-02 Bq 2.2 3.15E-03 Bq 1.6 9.34E-03 Bq 1.6

1972 4.43E-02 Bq 2.58 2.06E-03 Bq 1.69 6.57E-03 Bq 1.6

1973 1.37E-01 Bq 1.6 8.56E-03 Bq 1.6 3.87E-02 Bq 1.6

1974 5.24E-01 Bq 1.6 2.98E-02 Bq 1.6 8.28E-02 Bq 1.6

1975 2.64E-01 Bq 1.6 1.55E-02 Bq 1.6 4.97E-02 Bq 1.6

1976 8.37E-02 Bq 1.86 5.33E-03 Bq 1.6 1.56E-02 Bq 1.6

1977 4.61E-02 Bq 2.55 2.18E-03 Bq 1.7 4.79E-03 Bq 1.6

1978 3.84E-02 Bq 2.79 1.75E-03 Bq 1.84 3.68E-03 Bq 1.67

1979 3.72E-02 Bq 2.83 2.73E-03 Bq 1.6 6.06E-03 Bq 1.6

1980 6.29E-02 Bq 2 4.34E-03 Bq 1.6 1.64E-02 Bq 1.6

1981 4.20E-02 Bq 2.62 3.19E-03 Bq 1.6 9.16E-03 Bq 1.6

1982 3.32E-02 Bq 2.98 2.89E-03 Bq 1.6 9.93E-03 Bq 1.6

1983 3.41E-02 Bq 2.96 1.52E-03 Bq 1.95 2.50E-03 Bq 1.93

1984 3.40E-02 Bq 2.97 1.51E-03 Bq 1.95 2.46E-03 Bq 1.94

1985 3.39E-02 Bq 2.97 1.51E-03 Bq 1.95 2.45E-03 Bq 1.94

1986 3.74E-04 Bq 1.6 1.99E-05 Bq 1.6 2.45E-05 Bq 1.6

1987 6.65E-05 Bq 1.6 3.30E-06 Bq 1.6 5.57E-05 Bq 1.6

1988 2.11E-04 Bq 1.6 9.84E-06 Bq 1.6 2.39E-04 Bq 1.6

1989 1.12E-05 Bq 1.6 7.14E-08 Bq 1.6 1.56E-04 Bq 1.6

1990 1.72E-04 Bq 1.6 7.02E-06 Bq 1.6 2.82E-04 Bq 1.6

1991 4.23E-03 Bq 1.6 1.87E-04 Bq 1.6 5.63E-03 Bq 1.6

1992 1.06E-02 Bq 1.6 4.51E-04 Bq 1.6 1.57E-02 Bq 1.6

1993 2.39E-03 Bq 1.6 1.22E-04 Bq 1.6 1.67E-03 Bq 1.6

1994 1.86E-03 Bq 1.6 9.43E-05 Bq 1.6 1.39E-03 Bq 1.6

1995 5.17E-04 Bq 1.6 4.95E-06 Bq 1.6 2.28E-03 Bq 1.6
Effective Date: 12/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-4 Page 20 of 21

Table 4D-2. Plutonium annual intake per year (Bq) and geometric standard deviation. 

Date

Pu-238 Pu-239 Pu-240 Pu-241 Pu-242

50th

percentile GSD

50th

percentile GSD

50th

percentile GSD

50th

percentile GSD

50th

percentile GSD

1944

1945

1946

1947

1948

1949

1950

1951

1952

1953 3.12E-06 Bq 1.16 2.13E-05 Bq 1.16 5.07E-06 Bq 1.16 1.47E-04 Bq 1.16 7.17E-10 Bq 1.16

1954 1.37E-06 Bq 1.16 9.33E-06 Bq 1.16 2.22E-06 Bq 1.16 6.43E-05 Bq 1.16 3.14E-10 Bq 1.16

1955 1.42E-06 Bq 1.16 9.70E-06 Bq 1.16 2.31E-06 Bq 1.16 6.68E-05 Bq 1.16 3.26E-10 Bq 1.16

1956 6.67E-07 Bq 1.16 4.55E-06 Bq 1.16 1.08E-06 Bq 1.16 3.14E-05 Bq 1.16 1.53E-10 Bq 1.16

1957 6.81E-07 Bq 1.16 4.65E-06 Bq 1.16 1.11E-06 Bq 1.16 3.21E-05 Bq 1.16 1.56E-10 Bq 1.16

1958 3.81E-06 Bq 1.16 2.60E-05 Bq 1.16 6.18E-06 Bq 1.16 1.79E-04 Bq 1.16 8.73E-10 Bq 1.16

1959 1.09E-06 Bq 1.16 7.45E-06 Bq 1.16 1.77E-06 Bq 1.16 5.13E-05 Bq 1.16 2.50E-10 Bq 1.16

1960 2.04E-06 Bq 1.16 1.39E-05 Bq 1.16 3.31E-06 Bq 1.16 9.60E-05 Bq 1.16 4.68E-10 Bq 1.16

1961 1.51E-06 Bq 1.16 1.03E-05 Bq 1.16 2.46E-06 Bq 1.16 7.12E-05 Bq 1.16 3.47E-10 Bq 1.16

1962 3.81E-07 Bq 1.16 2.60E-06 Bq 1.16 6.18E-07 Bq 1.16 1.79E-05 Bq 1.16 8.74E-11 Bq 1.16

1963 2.00E-06 Bq 1.16 1.37E-05 Bq 1.16 3.25E-06 Bq 1.16 9.43E-05 Bq 1.16 4.60E-10 Bq 1.16

1964 9.37E-08 Bq 1.16 6.40E-07 Bq 1.16 1.52E-07 Bq 1.16 4.41E-06 Bq 1.16 2.15E-11 Bq 1.16

1965 5.26E-07 Bq 1.16 3.59E-06 Bq 1.16 8.54E-07 Bq 1.16 2.47E-05 Bq 1.16 1.21E-10 Bq 1.16

1966 6.74E-08 Bq 1.16 4.60E-07 Bq 1.16 1.09E-07 Bq 1.16 3.17E-06 Bq 1.16 1.55E-11 Bq 1.16

1967 6.30E-08 Bq 1.16 4.30E-07 Bq 1.16 1.02E-07 Bq 1.16 2.97E-06 Bq 1.16 1.45E-11 Bq 1.16

1968 8.93E-08 Bq 1.16 6.10E-07 Bq 1.16 1.45E-07 Bq 1.16 4.20E-06 Bq 1.16 2.05E-11 Bq 1.16

1969 1.18E-07 Bq 1.16 8.08E-07 Bq 1.16 1.92E-07 Bq 1.16 5.57E-06 Bq 1.16 2.71E-11 Bq 1.16

1970 9.27E-08 Bq 1.16 6.33E-07 Bq 1.16 1.51E-07 Bq 1.16 4.36E-06 Bq 1.16 2.13E-11 Bq 1.16

1971 1.38E-07 Bq 1.16 9.44E-07 Bq 1.16 2.24E-07 Bq 1.16 6.50E-06 Bq 1.16 3.17E-11 Bq 1.16

1972 1.48E-07 Bq 1.16 1.01E-06 Bq 1.16 2.41E-07 Bq 1.16 6.98E-06 Bq 1.16 3.40E-11 Bq 1.16

1973 2.73E-07 Bq 1.16 1.87E-06 Bq 1.16 4.44E-07 Bq 1.16 1.29E-05 Bq 1.16 6.27E-11 Bq 1.16

1974 5.61E-07 Bq 1.16 3.83E-06 Bq 1.16 9.11E-07 Bq 1.16 2.64E-05 Bq 1.16 1.29E-10 Bq 1.16

1975 1.01E-07 Bq 1.16 6.93E-07 Bq 1.16 1.65E-07 Bq 1.16 4.77E-06 Bq 1.16 2.33E-11 Bq 1.16

1976 9.67E-08 Bq 1.16 6.61E-07 Bq 1.16 1.57E-07 Bq 1.16 4.55E-06 Bq 1.16 2.22E-11 Bq 1.16

1977 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1978 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1979 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1980 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1981 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1982 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1983 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1984 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1985 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1986 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1987 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1988 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1989 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1990 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1991 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1992 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1993 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1994 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16

1995 5.90E-08 Bq 1.16 4.03E-07 Bq 1.16 9.58E-08 Bq 1.16 2.78E-06 Bq 1.16 1.35E-11 Bq 1.16
Effective Date: 12/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0009-4 Page 21 of 21

Table 4D-3. Neptunium and technetium annual intake per year (Bq) and geometric standard 

deviation. 

Date

Np-237 Tc-99

50th percentile GSD 50th percentile GSD

1944

1945

1946

1947

1948

1949

1950

1951

1952

1953 3.88E-05 Bq 1.16 4.98E+00 Bq 1.16

1954 1.70E-05 Bq 1.16 4.98E+00 Bq 1.16

1955 1.76E-05 Bq 1.16 4.98E+00 Bq 1.16

1956 8.27E-06 Bq 1.16 4.98E+00 Bq 1.16

1957 8.45E-06 Bq 1.16 4.98E+00 Bq 1.16

1958 4.72E-05 Bq 1.16 4.98E+00 Bq 1.16

1959 1.35E-05 Bq 1.16 4.98E+00 Bq 1.16

1960 2.53E-05 Bq 1.16 4.98E+00 Bq 1.16

1961 1.88E-05 Bq 1.16 4.98E+00 Bq 1.16

1962 4.73E-06 Bq 1.16 4.98E+00 Bq 1.16

1963 2.49E-05 Bq 1.16 2.86E+00 Bq 1.16

1964 1.16E-06 Bq 1.16 2.86E+00 Bq 1.16

1965 6.53E-06 Bq 1.16 2.86E+00 Bq 1.16

1966 8.37E-07 Bq 1.16 2.86E+00 Bq 1.16

1967 7.82E-07 Bq 1.16 2.86E+00 Bq 1.16

1968 1.11E-06 Bq 1.16 2.86E+00 Bq 1.16

1969 1.47E-06 Bq 1.16 2.86E+00 Bq 1.16

1970 1.15E-06 Bq 1.16 2.86E+00 Bq 1.16

Date

Np-237 Tc-99

50th percentile GSD 50th percentile GSD

1971 1.72E-06 Bq 1.16 2.86E+00 Bq 1.16

1972 1.84E-06 Bq 1.16 3.11E+00 Bq 1.16

1973 3.39E-06 Bq 1.16 5.16E+00 Bq 1.16

1974 6.96E-06 Bq 1.16 7.14E+00 Bq 1.16

1975 1.26E-06 Bq 1.16 2.36E+00 Bq 1.16

1976 1.20E-06 Bq 1.16 2.84E+00 Bq 1.16

1977 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1978 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1979 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1980 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1981 7.32E-07 Bq 1.16 1.67E-02 Bq 1.16

1982 7.32E-07 Bq 1.16 1.05E-02 Bq 1.16

1983 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1984 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1985 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1986 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1987 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1988 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1989 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1990 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1991 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1992 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1993 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1994 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

1995 7.32E-07 Bq 1.16 4.18E-07 Bq 1.16

Table 4D-4. Background gamma radiation levels at two perimeter ORGDP stations. 

Date Average GSD

1944 9.8 uR/hr 1.08

1945 9.8 uR/hr 1.08

1946 9.8 uR/hr 1.08

1947 9.8 uR/hr 1.08

1948 9.8 uR/hr 1.08

1949 9.8 uR/hr 1.08

1950 9.8 uR/hr 1.08

1951 9.8 uR/hr 1.08

1952 9.8 uR/hr 1.08

1953 9.8 uR/hr 1.08

1954 9.8 uR/hr 1.08

1955 9.8 uR/hr 1.08

1956 9.8 uR/hr 1.08

1957 9.8 uR/hr 1.08

1958 9.8 uR/hr 1.08

1959 9.8 uR/hr 1.08

1960 9.8 uR/hr 1.08

1961 9.8 uR/hr 1.08

Date Average GSD

1962 9.8 uR/hr 1.08

1963 9.8 uR/hr 1.08

1964 9.8 uR/hr 1.08

1965 9.8 uR/hr 1.08

1966 9.8 uR/hr 1.08

1967 9.8 uR/hr 1.08

1968 9.8 uR/hr 1.08

1969 9.8 uR/hr 1.08

1970 9.8 uR/hr 1.08

1971 9.8 uR/hr 1.08

1972 9.8 uR/hr 1.08

1973 7.4 uR/hr 1.09

1974 7.6 uR/hr 1.09

1975 8.2 uR/hr 1.08

1976 10.5 uR/hr 1.07

1977 7.6 uR/hr 1.09

1978 8.8 uR/hr 1.08

1979 9.6 uR/hr 1.07

Date Average GSD

1980 7.7 uR/hr 1.09

1981 7.8 uR/hr 1.09

1982 10.2 uR/hr 1.07

1983 9.9 uR/hr 1.07

1984 8.2 uR/hr 1.08

1985 8.8 uR/hr 1.08

1986 9.8 uR/hr 1.08

1987 9.8 uR/hr 1.08

1988 9.8 uR/hr 1.08

1989 9.8 uR/hr 1.08

1990 9.8 uR/hr 1.08

1991 9.8 uR/hr 1.08

1992 9.8 uR/hr 1.08

1993 9.8 uR/hr 1.08

1994 9.8 uR/hr 1.08

1995 9.8 uR/hr 1.08

* Denotes data derived from empirical site measurements.